ORNL-TM-7207 - the Molten Salt Energy Technologies Web Site

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104 Step 9. Precipitate the cerium as Ce02 and recover the precipitate by Step 10. decantation and filtration. Discard a portion of the molten salt, which contains rare earth fission products, to waste stor- age. Return the remainder with the necessary makeup to step 8. Combine the Ce82 from step 9 with that from step 7 and treat these sollids with %%F and H2 to obtain CeF3 (plus some ThF4). Use this as the major part of the reagent for step 5. This process would have a number of disadvantages when compared with the reductive-extraction/metal-transfer process. Zirconium, cesium, ru- bidium, strontium, and barium would not be removed, though none of these is a major problem. Neptunium probably would not be removed, though am- ericium and curium may be. Iodine W Q U ~ be ~ removed either during. the fuel QXidatiOn Or SUbSe3qUent hydrOflUOrfnaeiQnS. SeleIliWll and tellUriUIT3 - a%- swing that they arrive at the processing plant -rnsnight be volatilized as elements or as fluorides during the fuel oxidation step (and they might cause a C O ~ ~ O S problem ~ Q ~ for the process). Heat generation by the fuel, even after a few days cooling time, would present problems, and the con- plex process would be difficult (possibly impossible) to engineer. At best, several days would be required to get a batch of DMSR fuel solvent through the process, though the fissile materials might be returned to the reactor with a 2-el holdup. An appreciable inventory of fuel material (but perhaps not more than 5% of reactor inventory) would be cooling and in the processing area. 4*Ie4 Salt replacement Even with no chemical removal of fission products, the neutron poi- soning effect Fn a BMSR does not begin to approach saturation until after about 15 years of power operation at a 75% capacity factor. Thus, if the fission-product inventory could be held at or below that corresponding to a 15-year level, a significant reduction in fueling requirements could be realized. The simplest way ts l imit the fission-product concentration in the salt is t~ discard a portion of the salt on a routine schedule and replace it wfth clean salt. with no refinement, salt discard would require replacewent sf the fissile material as well as the fertile cowponerat
and the ~ dvent (or carrier) salt and, therefo~e, would actually require a larger uranium supply than the 30-year once-through fuel cycle proposed for the reference IBmR concept. However, uranium is easily and effectively separated frsm the rest of the fuel mixture, so the denatured uranium could be removed and recycled %e: the reactor site with a minimum of ef- fort. Depending on the rate 06 salt replacement, this approach would significantly reduce the requirement for fissile uranium below that fsr the simple once-through cycle 4.2 Fuel Cyde Performance Of the alternate fuel cycles considered in this secti~n, the break- even breeder, if it were successfd, would provide for the best utiliza- tion of fissile fuel resources ~ ~ 3 % ) . ~f that system were started up on 20% e ~ r i ~ 2h 3 e % ~ ~ it ~ would probably require 700 to 1000 ~g of natu- ral U308 to provide the initial fuel loading for each 1 CX of electric generating capability. [The separative work to enrich this fuel to 20% 235U would be less than 1 million separative work units (SWU),] However, once provided, this fuel would continue to p~~dluce electricity in an arbitrarily long succession of power stations (or as long as fertile material was available). Thus, the effective resource requirement could be made arbitrarily small by averaging It Over iB large number Of plants. Even if the initial fuel charge were used in only one pliant, the resource requirement would be only 110 to 20% of that for an LWR with similar electric generating capability. The converter options with fuel processing provide other estimates of the potential performance of DMSRs with fissi~n-product cleanup (Table 29). The options, which were described earlier, my be sumarized as f~klows: Option Fuel cycle A B c D Initial load is 20% 235U; makeup fuel is 20% 235U Initial load is 20% 235U; makeup fuel is 33% 735U Initial load is 20% 23sU; annual discard of 1% of uranium inventory; makeup fuel is 20% 235U Initial load is 20% 235U; annual reenrichment of 2% of uranium inventory to denaturin limit or to one-lialf of prior 7 3 8 ~ content; makeup fuel is 20% 255C
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P iii CONTENTS ....................
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CONCEPTUAL DESIGN CHARACTERISTICS O
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3 The primary purpose of this study
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.. .... . w..w c Y ... . r 5 2. GEN
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ii.... :.: ..... .,. \ ..... %& . 3
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addition StathQR would be required.
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!I \.... . , . ~ .... .. . . / I I
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... .. ...... . . . $i 13 Table 1.
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15 Initial scoping studies showed t
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.. . E 6> I S N 0.45 k- 0.50 B --.
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:.:..y ,. . . w, E9 The code calcul
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h 3. Additional 238Y is added as re
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...z,y .. . c .A, Table 9. Actinide
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';x.w' .... Relative power distribu
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........ . . . :..i 27 a Inner 2.54
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29 Table 16. Effect of neutron spec
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ob E 2 a 4 5 6 7 8 9 10 11 12 13 14
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.. . . . . . . . . . 'a# 0.1 0.2 0.
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35 (Doppler effect), (2) changing t
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where Ci = relative delayed-neutron
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.... . . . . , c.y..p c "
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.... :...w . . . . . :, %. , dimens
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,. ::.y ~. . .... , 40 0 43 1 2 3 4
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.,.,. . . ., , .. . . . . . . . %..
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.... ... . ....... . .%.W 3.3.1.1 C
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49 Variation of fuel composition wi
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.. .*,p. . . . . . . . Table 25. Ph
Page 59 and 60: ;: c.... x. . w ..i 53 Table 26, St
Page 61 and 62: would be expected to have an equili
Page 63 and 64: .... . . . . :i ,.*,&s 57 Corrosion
Page 65 and 66: . ;;..,..I . . . . . . "LW 0 9 m -
Page 67 and 68: of the fissioned atom. ea ~arly ass
Page 69 and 70: 63 n e results of a program of solu
Page 71 and 72: 65 Table 27. Sources and rates Sf p
Page 73 and 74: operation of the PISRE. 67 analyses
Page 75 and 76: UP3/UF4 ratio would be possible. DM
Page 77 and 78: $ .p9 3.3.3.1 Preparation of initia
Page 79 and 80: 73 Contamination of fuel by seconda
Page 81 and 82: *w- through the walls of the primar
Page 83 and 84: 77 4 will apparently be necessary (
Page 85 and 86: such contamination. For a demonstra
Page 87 and 88: 8% 3.4 Reactor Materials Although s
Page 89 and 90: : ....., . .’ -. ..WY used in the
Page 91 and 92: ..... i.. ....., . .Yd 85 QRPIL-DWG
Page 93 and 94: ... w . . .y and to salt containing
Page 95 and 96: With the different ~Bjectives of ns
Page 97 and 98: t.;$- ... quite different from thos
Page 99 and 100: 93 during noma1 operation. Minor am
Page 101 and 102: ............. . . .. . .=w as SSm,
Page 103 and 104: . . . :.w . . . . , 4. ALTERNATIVE
Page 105 and 106: .. . ... . . . . . . .. . . "W 3. T
Page 107 and 108: .,.. w .." by removing some uranium
Page 109: , . . . . . . , -c%w probably still
Page 113 and 114: BO 7 TPle fuel makeup requirements
Page 115 and 116: 5. CQPIPIERCIWLPZATION CONSIDERATIO
Page 117 and 118: 11% At the close of MSRE operation,
Page 119 and 120: 12YS 75 22”” 955 30” 1””
Page 121 and 122: 115
Page 123 and 124: aa 7 some components in common with
Page 125 and 126: Ae eo un t No * Table 3%. Capital c
Page 127 and 128: - ,.,.,.,., p .. ..... . G is based
Page 129 and 130: 123 Table 34. Summary of annual non
Page 131 and 132: ...... . . . . ., x.:w* 4 i.. ... _
Page 133 and 134: E2 7 There are no detailed estimate
Page 135 and 136: 129 particularly if such activities
Page 137 and 138: . ~.........* . . . . . :*. ... . -
Page 139 and 140: . .. . . . . . . . ;.:w b 133 the i
Page 141 and 142: _... ..&/ ; .; with reactor scram m
Page 143 and 144: 137 includlng $900 million for the
Page 145 and 146: 139 The authors want to acknowledge
Page 147 and 148: . ... . .,.,.,.,.,. . , i.rc- q....
Page 149 and 150: .:. _.i .. ..... -.=* 45. A. L. Mat
Page 151 and 152: ...... &;.& 3 14 5 74. W, R. Grimes
Page 153 and 154: p 103. 6. Re Hudson 11, CQNCEPT-5 U
Page 155: ... . . . . . . . , . ,.Y&fl 149 ap
Page 158 and 159: Account No. -0- Three- digit digit
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.... ..
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156 77. He 6. MacPherson, Institute
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